In Dix et al. U.S. Pat. No. 5,112,570 issued May 12, 1992, entitled TWO-PHASE PRESSURE DROP REDUCTION BWR ASSEMBLY DESIGN (formerly U.S. patent application Ser. No. 07/176,975 filed Apr. 4, 1988), a fuel bundle having a plurality of part length rods was illustrated. A summary of that construction and the advantages set forth in this reference can be instructive.
Construction of the fuel bundle in Dix et al. is conventional with the exception of the addition of less than full length fuel rods. The conventional portion of the disclosed fuel assembly in the Dix et al. Patent is easy to understand. This assembly includes a channel having vertically extending walls for extending around a fuel bundle assembly volume. The channel is open at the bottom for receipt of water moderator and open at the top for the discharge of water and steam. The fuel bundle includes a matrix of vertically upstanding fuel rods--these rods being sealed tubes containing fissionable materials. The fuel rods are supported on a lower tie plate which permits the entry of the water moderator to the fuel bundle. The fuel rods typically extend to an upper tie plate which maintains the fuel rods in their side by side vertical relation and permits the generated steam and remaining water to escape.
The Dix et al. disclosure adds to the conventional fuel assembly, a plurality of less than full length spaced apart so-called "part length (fuel) rods" (PLRs). These fuel rods are supported on the lower tie plate, extend upwardly to and toward the upper tie plate, but terminate short of the upper tie plate. Between the point of part length fuel rod termination and the upper tie plate, the part length fuel rod defines in the upper two phase region of the fuel bundle a vent volume. This vent volume preferentially receives vapor from the liquid vapor two phase mixture in the upper two phase region of the fuel bundle during power producing operation.
Numerous advantages result from the part length rod construction. Improved cold shut down margin enables fuel to be designed with reduced amounts of burnable absorbers such as gadolinium. The tendency of the fuel bundle in the reactor to produce plutonium at the top of the bundle from resonance neutron capture in uranium 238 is reduced. The void overlying the part length rod has an increased vapor fraction with the result that the full length rods adjacent the voids have an increased liquid fraction. Further, the pressure drop in the upper two phase region of the fuel bundle is reduced. This being the case, the fuel bundle enjoys increased stability from thermal hydraulic and nuclear instabilities.
The fuel bundles are elongate. Further, the fuel rods contained within the fuel bundle are flexible. These fuel rods can flex out of their designed side-by-side spacing--and even into interfering contact with one another--due to flow induced vibration and rod bow. Therefore, spacers are utilized throughout the length of the fuel bundle.
Fuel bundle spacers have the function of maintaining the individual fuel rods at given elevations in their designed side-by-side relationship. Such spacers usually define a matrix of individual fuel rod containing cells. These cells fit around each and every fuel rod at their particular elevation in a fuel bundle. The fuel bundle spacers maintain the fuel rods in their designed side-by-side relationship and prevent interfering contact between the individual fuel rods. In the case of the part length rods where the fuel rods do not extend to the upper tie plate, the spacers maintain the fuel rods in their designed upstanding relation.
All fuel bundles--including those having part length rods--must be designed to operate within thermal limits. Specifically, that thermal limit in boiling water reactors known as critical power has always been a limitation. Critical power originates from rupture of the coolant liquid film on the exterior surface of the fuel rod in a phenomena known as "transition boiling."
In this transition boiling condition a liquid film no longer coats the exterior surface of the fuel rod. The rod on the exterior surface is exposed to coolant vapor only. Heat transfer from the fuel interior of the fuel rod undergoing fission reaction to the coolant is reduced. The fuel rod cladding becomes overheated. Naturally, as any fuel rod within a fuel bundle even approaches such a boiling condition anywhere along its length, power is restricted to avoid violation of this "critical power" limitation.
Past experimentation has been directed to the critical power limitation. It is known that by decreasing the spacer pitch in the upper two phase region of the fuel bundle, that critical power can be improved. Unfortunately, the additional spacers caused additional pressure loss. This additional pressure loss causes additional tendencies for instabilities at certain power rates of the reactor. These instabilities include local and core wide thermal hydraulic and nuclear thermal hydraulic instabilities. For these reasons, the experimentally determined improvement of critical power could not be implemented by decreasing the spacer pitch in the upper two phase region of boiling water nuclear reactor fuel bundles.
It is also known to incorporate so-called "swirl vanes" to both boiling water nuclear reactors and the spacers in boiling water nuclear reactors. These devices can be simply summarized and easily understood.
In summary, so-called swirl vanes are placed interstitially of fuel rods. The vanes themselves comprise pieces of metal twisted in a helical pattern. In the earliest known cases, these so-called swirl vanes were the same length as the fuel rods in the reactor. In a later case, a spacer constructed from such swirl vanes was constructed. See Johansson, U.S. Pat. No. 4,913,895 issued Apr. 3, 1990 entitled SWIRL VANES INTEGRAL WITH SPACER GRID.
These swirl vanes when added to reactors had a beneficial effect and a detrimental effect.
The beneficial effect was the classification of water from upwardly flowing water and steam. Specifically, upwardly flowing water and steam. Simply stated, and despite the helical pattern of the twisted metal strips, steam tended to upwardly flow about the swirl vanes. Water, however, did not tend to join this upward flow. Instead the heavier water received a horizontal velocity component from the swirl vanes. As the swirl vanes were placed interstitially of the fuel rods, the heavier water when thrown horizontally by the momentum of the swirl vanes has the beneficial effect of impacting the adjacent fuel rods. Consequently, the critical power limit is increased.
The detrimental effect of such swirl vanes is increased pressure drop. The swirl vanes themselves raise the pressure drop in the upper two phase region of the boiling water reactor. This increase in pressure drop will increase the possibility of instabilities including thermal hydraulic instabilities and nuclear, thermal hydraulic instabilities at high power/low flow conditions of the boiling water nuclear reactor. This being the case, the swirl vanes have not been in large measure introduced into the boiling water nuclear reactors.
Any physical explanation of spacer relative thermal hydraulic performance should depend on the flow regimes that the coolant experiences in flowing up the channel as well as how the flow interacts with the spacer.
Single phase water enters the bottom of the fuel assembly and is heated until sub cooled boiling occurs. Bubbles are formed at the surface of the fuel rod but quickly condense as they contact the bulk sub cooled flow. At the 100% power/100% flow condition bundle average bulk boiling will begin somewhere between the bottom spacer and the second spacer from the bottom of the fuel assembly. Now bubbles in the main flow stream will grow and the flow regime will progress from bubble flow to a type of slug or froth flow where individual small bubbles are starting to combine to make larger slugs of vapor. During these processes the vapor is flowing as bubbles or slugs in a continuous liquid medium.
Depending on conditions somewhere around the middle of the bundle a flow regime transformation takes place. Now there is so much vapor that it becomes the continuous medium and the liquid is either found as a thin film flowing on all the solid surfaces of the bundle or as droplets entrained in the continuous vapor. This is the annular flow regime which is important because it is where dry out or boiling transition will commonly take place in a BWR.
The limiting critical power condition in a BWR has been referred to in the literature alternately as dry out, boiling crisis, critical heat flux, burnout and boiling transition, the term which will be used here. Boiling transition is defined as the first condition of degraded heat transfer in the fuel bundle. This occurs in the annular flow regime as a result of the thin liquid film which covers all the fuel rod surfaces going to zero film thickness. A critical power problem results.